ML17219A440

Amend 18 to License NPF-16,revising Pressure/Temp Limits for Steam Generators & Making Several Administrative/Editorial Changes

ML17219A440
Person / Time
Site:	Saint Lucie
Issue date:	03/11/1987
From:	Thadani A Office of Nuclear Reactor Regulation
To:	City of Orlando, FL, Florida Municipal Power Agency, Florida Power & Light Co, Orlando Utilities Commission
Shared Package
ML17219A441	List: ML17219A440 ML17219A442
References
NPF-16-A-018 NUDOCS 8703190496
Download: ML17219A440 (14)

Text

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+c +y*y4 UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 FLORIDA POWER 8( LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.

LUCIE PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

18 License No.

NPF-16 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Florida Power 8 Light Company, et al. (the licensee),

dated June 17, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment wi 11 not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

H, 05000~8+

496 870311 P

2.

Accordingly, Facility Operating License No.

NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.2 to read as follows:

2.

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

18, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 11, 1987 FOR THE NUCLEAR REGULATORY COMMISSION

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Ashok

. Tha anl,

>rector PWR Project Directorate ¹8 Division of PWR Licensing-B

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ATTACHMENT TO LICENSE AMENDMENT NO. 18

\\

TO FACILITY OPERATING LICENSE NO.

NPF-16 DOCKET NO. 50-389 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Pa es 3/4 4-1 3/4 7"12 3/4 7-30 B 3/4 7-4 6-18 Insert Pa es 3/4 4-1 3/4 7"12 3/4 7-30 B 3/4 7-4 6-18

3/4.4 REACTOR COOLANT SYSTEM 3/4.4;1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4. 1.1 Both Reactor Coolant loops and both Reactor Coolant pumps in each loop shall be in operation.

APPLICABILITY: 1 and 2."

ACTION:

With less than the above required Reactor Coolant pumps in operation, be in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SURVEILLANCE RE UIREMENTS

4. 4. l. 1 The above required Reactor Coolant loops shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• See Special Test Exception 3.10.3.

ST.

LUCIE - UNIT 2 3/4 4"1 Amendment No.

18

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4. 1.2 The Reactor Coolant loops listed below shall be OPERABLE and at least one of these Reactor Coolant loops shall be in operation."

a.

Reactor Coolant Loop 2A and its associated steam generator and at least one associated Reactor Coolant pump.

b.

Reactor Coolant Loop 2B and its associated steam generator and at least one associated Reactor Coolant pump.

APPLICABILITY'ODE3*"

ACTION:

With less than the above. required Reactor Coolant loops

OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the-next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With no Reactor Coolant loop in operatio'n, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Reactor Coolant loop to operation.

SURVEILLANCE RE UIREMENTS 4.4. l. 2. 1 At least the above required Reactor Coolant. pumps, if not..in operation, shall be determined to be OPERABLE=once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4. 1.2.2 At least one Reactor Coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4. 1.2.3 The required steam generator(s) shall be determined OPERABLE verifying the secondary side water level to be

> 10K indicated narrow range level at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

All Reactor Coolant pumps may be deener gized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10~F below saturation temperature.

""The requirements of Specification 3.4. 1.2 may be suspended for natural circulation training prior to initial'criticality.

ST.

LUCIE " UNIT 2 3/4 4-2

PLANT SYSTEMS ATMOSPHERIC DUMP VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7

. The atmospheric dump and associated block valves shall be OPERABLE

~ with:

a ~

b.

All atmospheric dump valves in manual control above 15K of RATED THERMAL POWER, and No more than one atmospheric dump valve per steam generator in automatic control below 15K of RATED THERMAL POWER.

APPLICABILITY:

MODE 1.

ACTION:

a 0 b.

With less than one atmospheric dump and associated block valve per steam generator OPERABLE; restore the required atmospheric dump and associated block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With more than the permissible number of atmospheric dump valves in automatic control, return the atmospheric dump valves to manual control within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4',.7.1.7,Each. atmospheric dump valve shall be verified to be in the manual

~operation mode at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during operation at

> 15K of RATED THERMAL POWER.

ST.

LUCIE UNIT 2 3/4 7-11

PLANT SYSTEMS 3 4.7.2 STEAM GENERATOR PRESSURE TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperatureof the secondary coolant in the steam generators shall

" be greater than 100'F when the pressure of the secondary coolant in the steam

'generator is greater than 200 psig.

APPLICABILITY: At all times.

ACTION'ith the requirements of the above specification not satisfied:

a.

Reduce the steam generator pressure to less than or equal to 200 psig wi.thin 30 minutes, and b.

Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator:

Determine that 'the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200'F.

SURVEILLANCE RE UIREMENTS 4.7.2 The pressure of the secondary side of the steam generators shall be determined to be less than 200'psig at least once per hour when the temperature of the secondary coolant is less than 100'F.

ST.

LUCIE - UNIT 2 3/4 7-12 Amendment No. 18

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued b.

Stored sources not in use - Each sealed source and fission detector-shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months.

Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

C.

Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source or detector.

4. 7. 10. 3 Reports. - A report shall be prepared and submitted to the Commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.

ST.

LUCIE - UNIT 2 3/4 7-29

PLANT SYSTEMS 3/4.7;ll FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM" LIMITING CONDITION FOR OPERATION 3.7. 11.1-The fire suppression water system shall be OPERABLE with:

a.

Two fire suppression

pumps, each with a capacity of 2350 gpm, with their discharge aligned to the fire suppression

header, b.

Separate water supplies, each with a minimum contained volume of 300,000 gallons, and c.

An OPERABLE flow path capable of taking suction from the city water storage tank lA and the city water storage tank 1B and transferring the water,=-through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrants, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe required to be OPERABLE per Specifications 3..7. 11.2, 3.7.11. 3 and 3.7.11. 4.

APPLICABILITY: At all times.

ACTION:

'a 0 With, one pump and/or one water supply inoperable, restore the inoperable equipment to OPERABLE status. within 7 days or prepare and submit a Special Report to the Commission pursuant to Speci-fication 6.9.2 within the next 30 days outlining the plans and procedures to be used to restore the inoperable equipment to OPERABLE,status, or.to provide;an,.alternate

~backup pump or supply.

The provisions of Specifications 3.0.3 and 3.0.'4 are not applicable.

b.

With the fire suppression water system otherwise inoperable, establish a backup fire suppression water system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This system is shared between St.

Lucie Units 1 and 2.

ST.

LUCIE - UNIT 2 3/4 7"30 Amendment No. pp, l8

.PLANT SYSTEMS BASES 3/4. 7. 1.4.

ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR

, Part 100 limits in the event of a steam line rupture.

This dose also includes the effects of a coincident 1.0 gpm primary to.secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power.

These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line

.. rupture.

This restriction is required to (1) minimize-the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation

'alves within the closure times of the Surveillance Requirements is consistent with the assumptions used in the safety analyses.

3/4. 7. 1. 6 MAIN.FEEDWATER LINE ISOLATION VALVES The main feedwater line isolation valves are required to be OPERABLE to ensure that (1) feedwater is terminated to the affected steam generator following

.-'a steam line break and (2) auxiliary feedwater is delivered to the intact steam generator following a feedwater line break.

If feedwater is not terminated to a steam generator with a. broken.main steam line, two serious effects may result:

(1) the post-trip return to power due to plant cooldown will be greater with resultant higher fuel failure and (2) the steam released to containment will exceed the design.

Due to removal of the main feed check valve from the plant design and its replacement with a second main feedwater line isolation valve, there is nothing other than the main feedwater line isolation valves to prevent back flow of AFW following a feed line break.

This may result in a loss of condensate inventory and the potential for not being able to feed the steam generator.

The concern is the failure of one main feedwater line isolation valve to close with the other main feedwater line isolation valve in that line being inoperable (i.e., stuck open).

It is thus desired to preclude operation for extended periods with a main feedwater line isolation valve known to be stuck in the open position.

ST.

LUCIE " UNIT 2 8 3/4 7-3

PLANT SYSTEMS BASES 3/4.'7.1.7 ATMOSPHERIC OUMP VALVES The limitation on maintaining the atmospheric dump valves in the manual mode of operation is to ensure the atmospheric dump valves will be'closed in the event of a steam line break.

For the steam line break with atmospheric dump valve control failure-event, the failure of the atmospheric dump valves to close would be a valid concern were the system to be in the automatic mode during power operations.

3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION The limitation.on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum al.lowable fracture toughness stress limits.

The limitations to 100 F and

.200 psig are based on a steam generator RTN>T of 'pO F and are sufficient to prevent brittle fracture.

3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that sufficient cooling capacity is available for;continued operation of'afety-related equipment during normal and accident conditions.

The redundant cooling capacity.of this

'system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

3/4.7.4 INTAKE COOLING WATER SYSTEM The OPERABILITY of the Intake Cooling Water System ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions.

The redundant cooling capacity of this

system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

ST.

LUCIE - UNIT 2 B 3/4 7-4 Amendment No. l8

ADMINISTRATIVE CONTROLS ANNUAL'EPORTS (Continued) greater than 100 mrems/yr and their associated man-rem exposure according to work and job functions, e. g., reactor operations and 2/

surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20K of the individual total dose need not be accounted for.

In the aggregate, at least 80K of the total whole cbody dose received from external sources should be assigned to specific major work functions.

MONTHLY OPERATING REPORTS 6.9. 1.6 Routine reports of operating statistics and shutdown experience,

.including documentation of all challenges to the..PORVs or safety valves, shall

- be submitted on a monthly basis to the Director, Office of Resource Management, U.S.

Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Administrator of the Regional Office of the

NRC, no later than the 15th of each month following the calendar month covered by the report.

This tabulation supplements the requirements of f20.407 of 10 CFR Part 20.

2 ST.

LUCIE " UNIT 2 6-17 Amendment No.

13

ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT 'RELEASE REPORT"

'6.9.1.7 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and.July 1 of each year.

The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide l.'21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials-in Liquid and Gaseous Effluents form Light-Water-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year.

This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if.measured),

or in the form of joint frequency distributions of wind speed,-wind direction, and atmos-pheric stability.""

This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit, or station during the previous calendar year.

This same report shall also include an assessment of the radiation doses from radioactive

- liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities

'nside, the SITE BOUNDARY (Figure 5. 1-1) during the report period.

All'ssump-tions used in making,,these assessments, ice

, specific activity, exposure time and location, shal'1 be included in these reports.

The meteorological condi-tions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses.

The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

= Every 2*years. using. the previous 6 months~release history for isotopes and historical meteorolgical data determine the controlling age group for both liquid and gaseous pathways.

If changed from current submit change to ODCM to

. reflect new tables for these groups and use the new groups in subsequent dose calculations.

The Radioactive Effluent Release Report to be submitted 60 days after January 1

of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases for the previous A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste

systems, the submittal shall specify the, releases of radioactive material 'from each unit.

In lieu of submission with the Radioactive Effluent Release

Report, the

.-licensee has the option of retaining this summary of required, meteorological data on site in a file that shall be provided to the NRC upon request.

ST.

LUGIE " UNIT 2 6-18 Amendment No.

18 18